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Add PNNL material composition library (#4134)
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‎docs/source/usersguide/materials.rst‎

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@@ -75,13 +75,87 @@ argument `enrichment_type`. For example the following would enrich Li6 to 15wo%:
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mat.add_element('Li', 1.0, enrichment=15.0, enrichment_target='Li6',
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enrichment_type='wo')
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Often, cross section libraries don't actually have all naturally-occurring
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isotopes for a given element. For example, in ENDF/B-VII.1, cross section
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evaluations are given for O16 and O17 but not for O18. If OpenMC is aware of
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what cross sections you will be using (through the
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:envvar:`OPENMC_CROSS_SECTIONS` environment variable), it will attempt to only
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put isotopes in your model for which you have cross section data. In the case of
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oxygen in ENDF/B-VII.1, the abundance of O18 would end up being lumped with O16.
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Some older cross section libraries do not have all naturally-occurring isotopes
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for a given element. For example, in ENDF/B-VII.1, cross section evaluations are
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given for O16 and O17 but not for O18. If OpenMC is aware of what cross sections
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you will be using through ``openmc.config['cross_sections']``, it will attempt
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to only put isotopes in your model for which you have cross section data. In the
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case of oxygen in ENDF/B-VII.1, the abundance of O18 would end up being lumped
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with O16.
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------------------
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Material Libraries
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------------------
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OpenMC includes material compositions from Revision 2 of the `PNNL Compendium of
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Material Composition Data for Radiation Transport Modeling
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<https://doi.org/10.2172/1782721>`_. A material can be created from this library
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with :meth:`Material.from_library`, for example::
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sodium_oxide = openmc.Material.from_library('Sodium Oxide')
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Keyword arguments accepted by :class:`Material` can also be specified, such as
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a material ID or temperature::
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salt_water = openmc.Material.from_library(
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'Salt Water', material_id=10, temperature=293.15)
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Material names are case sensitive, but whitespace is normalized when looking up
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a name. The default bundled library is ``'pnnl_v2'``. Each call returns a new,
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independent material with the density and composition reported in the compendium.
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The available material names can be retrieved without creating any materials::
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names = openmc.Material.get_library_material_names()
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Natural compositions in the PNNL library are stored as elemental atom fractions
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and expanded using :meth:`Material.add_element`. As described above, OpenMC will
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account for the nuclides available in the cross section library indicated by
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``openmc.config['cross_sections']``. Materials that are explicitly isotopic in
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the compendium, such as enriched uranium or He-3 proportional gas, retain those
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specified nuclides. OpenMC cannot substitute for a required isotope if it is
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absent from the selected nuclear data library.
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The compendium values are representative material definitions; actual density
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and composition can vary. Users should confirm that a library material is
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appropriate for their application.
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Custom libraries can be registered from local JSON files. Registration applies
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to the current Python process and cannot replace a library with the same name::
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openmc.Material.register_library('custom', 'custom_materials.json')
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names = openmc.Material.get_library_material_names('custom')
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material = openmc.Material.from_library(
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'My Material', library='custom', temperature=600.0)
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Custom files use the same schema as the bundled library. A minimal library
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containing an elemental composition is shown below:
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.. code-block:: json
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{
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"schema_version": 1,
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"density_units": "g/cm3",
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"percent_type": "ao",
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"materials": {
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"My Material": {
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"density": 1.0,
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"elements": {
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"H": 0.666667,
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"O": 0.333333
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}
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}
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}
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}
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Each material can contain ``elements``, ``nuclides``, or both. Elemental
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components are expanded using :meth:`Material.add_element` and
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``openmc.config['cross_sections']``. Nuclide components retain their explicitly
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specified isotopic composition. The top-level ``percent_type`` and
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``density_units`` values are passed to :class:`Material`, which validates them
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along with the density and components when a material is created. Component
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values can be relative amounts and do not need to sum to one. Additional
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provenance metadata, such as a top-level ``source`` or a per-material note, is
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permitted and ignored when constructing a material.
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-----------------------
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Thermal Scattering Data
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materials.cross_sections = '/path/to/cross_sections.xml'
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.. _MCNP: https://mcnp.lanl.gov/
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